Monte Carlo optimization of an epithermal neutron flux monitor for BNCT

An epithermal neutron (0.5 eV < E_n < 10 keV) flux monitor developed for boron neutron capture therapy (BNCT) was optimized by Monte Carlo simulations. Based on this optimization study, the optimization results for each component of the epithermal neutron flux monitor were obtained. The simulation results indicated that the epithermal neutron flux monitor with optimal configuration was more efficiently applicable to precisely measure the epithermal neutron fluxes of BNCT neutron sources.     

中子反射谱仪前端导管优化

基于两种中子反射谱仪设计(A:前端为8.0 m垂直聚焦导管,入口200mm,出口40 mm;B:前端为1.6+3.4m水平偏转垂直聚焦导管,入口100mm,出口50mm),对其前端偏转/聚焦导管进行了中子束模拟,结果表明:1)在特征波长为0.4 nm的冷中子导管末端,没有必要再次偏转束流以减低本底;2)与建议方案相比,概念设计中的注量下降了52.8%,而水平方向发散度增大了24.1%,垂直方向发散度减小60.7%;3)聚焦导管长度应不低于6.0 m,导管聚焦超镜m值在2.0-2.5最佳,两侧超镜m值对束流几乎无影响,使用中子束高度应不低于150mm可达到较好的注量率和发散度.     

Design of an epi-thermal neutron flux intensity monitor with GaN wafer for boron neutron capture therapy

Boron neutron capture therapy (BNCT) is a promising cancer therapy. Epi-thermal neutron (0.5 eV < E_n < 10 keV) flux intensity is one of the basic characteristics for modern BNCT. In this work, based on the ⁷¹Ga(n,γ)⁷²Ga reaction, a new simple monitor with gallium nitride (GaN) wafer as activation material was designed by Monte Carlo simulations to precisely measure the absolute integral flux intensity of epi-thermal neutrons especially for practical BNCT. In the monitor, a GaN wafer was positioned in the center of a polyethylene sphere as neutron moderator covered with cadmium (Cd) layer as thermal neutron absorber outside. The simulation results and related analysis indicated that the epi-thermal neutron flux intensity could be precisely measured by the presently designed monitor.     

Design study of neutron flux intensity monitor between ten and several hundred keV for BNCT

Based on the activation method using ⁷¹Ga(n,γ)⁷²Ga reaction, two spherical monitors with gallium nitride (GaN) wafers as activation material were designed by Monte Carlo simulations to precisely measure the absolute integral neutron flux intensity between ten and several hundred keV. The two monitors are almost the same in shape and have an absorber/moderator/absorber/GaN arrangement from outside to inside. The differences between the two monitors are the kind of materials, the thicknesses of the absorbers and the diameter of the moderator. By making difference of the sensitivities between these two monitors, the contributions of thermal, epithermal and very high energy fast neutrons were removed completely, and constant monitor sensitivity to neutrons between ten and several hundred keV was extracted. The simulation results and related analysis indicated that the absolute integral neutron flux intensity between ten and several hundred keV could be precisely measured by the presently designed two monitors.     

Neutron flux characterization of the SM1 sub-critical multiplying complex of the Pavia University

SM1 is a thermal Sub-critical Multiplication complex located at the University of Pavia (Italy) and, since its installation in 1962, has been utilized mainly for radiochemistry research. This work focuses on the characterization, by means of the Monte Carlo code MCNP and direct measurements, of the neutron flux distribution inside the complex and on the calculation of the effective multiplicative coefficient (k_eff) in the current SM1 thermal configuration. For two specific irradiation channels, experimental measurements of the neutron fluxes were performed by foils activation technique and neutron spectrum de-convolution based on the SAND II code. Measurements have been compared with the simulation results showing a good agreement. Furthermore, a comparison between the preliminary results of the simulations of the SM1 plant in fast configuration, characterized by a solid lead diffuser, and the actual thermal configuration is also presented. The fast configuration of SM1, if implemented, will give the opportunity to carry out preliminary studies for the analysis of sub-critical fast-neutron installations and their applications.     

Simulation and experimental study of a random neutron analyzing system with （252）~Cf neutron source

Experiments were performed on a high-speed online random neutron analyzing system(HORNA system)with a 252Cf neutron source(up to 1 GHz sampling rate and 3 input data channel),to obtain time-and frequencydependent signatures which are sensitive to changes in the composition,fissile mass and configuration of the fissile assembly.The data were acquired by three high-speed synchronized acquisition cards at different detector angles,source-detector distances and block sizes.According to the relationship between 252Cf source and the ratio of power spectral density,Rpsd,all the signatures were calculated and analyzed using correlation and periodogram methods.Based on the results,the simulated autocorrelation functions were utilized for identifying different fissile mass with Elman neural network.The experimental results show that the Rpsd almost remains at constant amplitude in frequency range of 0-100 MHz,and is only related to the angle and source-detector distance.The trained Elman neural network is able to distinguish the characteristics of autocorrelation function and identify different fissile mass.The average identification rate reached 90% with high robustness.     

CARR上的应力测量中子衍射谱仪概念设计和模拟研究

中子衍射法是迄今为止可直接测量材料或工程部件内部深处应力场分布的唯一非破坏性方法，在工程上具有重要的应用。中国先进研究堆（CARR)中子散射工程拟建造一台应力测量中子衍射谱仪，其主要功能是测量材料中的残余应力和载荷应力。本文介绍了该谱仪的概念设计方案，并应用蒙特卡罗模拟软件MCSTAS对设计方案进行了模拟研究，对部分中子部件参数进行了优化设计。     

用于CSNS反角白光中子能谱及注量率测量的快裂变室

即将建成的中国散裂中子源(China Spallation Neutron Source,CSNS)反角白光中子束线可为核数据测量提供高注量率的脉冲白光中子束流,填补我国核数据测量用白光中子源的空白,提高我国核数据测量水平,满足核能、核技术及基础核物理研究对核数据的需求。该束线建成后,其中子能谱及注量率的精确测量将是开展其它物理实验的基础,快裂变电离室因其独特优点被选为中子能谱和注量率测量探测器。通过实验研究了快裂变电离室的粒子分辨性能、时间分辨性能;确定阴、阳极的合理间距为10 mm,据此测得电离室的时间分辨约15 ns;利用235U样品量计算的探测效率与利用伴随粒子法给出的探测效率在不确定度范围内符合,因此可以标定快裂变室的探测效率。通过这些工作,完成了满足反角白光中子束能谱及注量率测量需求的快裂变室的物理设计。     

Study of relation between the gamma flux buildup factors and source geometry by M-C simulation

This paper is to achieve a gamma-ray source with the lowest rate of buildup factor, which is of great importance in medical, industrial and agricultural sciences.The flux buildup factor of gamma rays is calculated by the MCNP code for point, linear, surface and volume sources with shield layers of lead, iron and aluminum. The results show that for the high Z shielding material, the flux buildup factor of coaxial cylindrical sources is the lowest(1.6–2.3)of all sources, while for low Z shielding materials, the coaxial disk surface sources have smaller buildup factor(1.45–1.6).     

Measurement and calculation of neutron densities in BWR high burn-up fuels

A neutron-scanning device was developed for measuring accurate neutron densities of BWR high burn-up fuels up to 65 GWd tU⁻¹. Characteristic test of this device was done with a ²⁵²Cf source and adopted to measure axial distributions of neutron densities of BWR spent fuels with various enrichments (2.0–3.4%), which had been irradiated up to 60 GWd tU⁻¹ at Fukushima Daini Nuclear Power Station Unit 2(2F-2). We found the measured neutron densities were proportional to about fourth power of the corresponding burn-up values. The neutron densities calculated by the ORIGEN2.1 code and various cross section libraries showed good agreements with the measured ones in profile and absolute value except for BWR-UE file mainly based on ENDF/B-IV. The BS240J32 library based on JENDL3.2 was the best among the investigated libraries.     

材料与构件深部应力场及缺陷无损探测中子谱仪热中子导管概念设计

针对中国先进研究堆（CARR）正在建造的材料与构件深部应力场及缺陷无损探测中子谱仪所需的热中子导管,开展模拟计算与概念设计。首先根据CARR内的现场情况和该谱仪的整体要求设计热中子导管的内部截面尺寸为90 mm×160 mm,整体长度为19.7 m,导管长度分为3组;然后根据这些参数开展蒙特卡罗模拟,通过比较导管镀层的特征增殖因数m分别为1、2、3、4、5、6时导管末端的中子强度二维空间分布、水平方向发散角分布、波长分布等主要性能指标的模拟结果,选定m=3,并以此完成了导管的参数设计。     

Primary study on holdup measurement of ^235U in pipe using γ-ray spectrometry and Monte Carlo simulation

The pipe holdup measurement is very important for decommissioning nuclear facilities and nuclear-material control and accounting. The absolute detection efficiencies （εsp） of full-energy γ rays peak under different source density distribution function have been simulated using the Monte Carlo （MC） software, and the counting rates （no） of the characteristic γ rays have been measured using the γ spectrometer followed by the calculation of the holdup. The holdup is affected by the energy of γ rays, distance at which they are detected, pipe material, thickness, and source distribution of pipe, especially source distribution at a short distance. The comparative test of ^235U reference materials on the inner wall of Fe and A1 pipes （the total mass of ^235U is 44.6 mg and 222.8 mg, respectively） have been accomplished using this method. The determined result of ^235U is 43.2mg （U0.95rel=5.4%） and 216.2mg （U0.95rel= 3.2%）, respectively, which are in accordance with the reference values.     

The range of penetration and the backscattering coefficient by using both the analytic and the stochastic theoretical ways of electron slowing down in solid targets: Comparative study

The aim of this paper is to determine the best theoretical way (stochastic or analytic) to use in the study of the range of penetration and the backscattering coefficient of electron impinging in solid target, by adopting the same input in the form of collision cross-sections. For this purpose, differential elastic cross sections has been calculated by using our semi-empirical model [A. Bentabet, Z. Chaoui, A. Aydin, A. Azbouche, Vacuum 85 (2010) 156]; and that of inelastic cross sections has been calculated by using Gryzinski’s excitation function. Moreover, in stochastic case, the obtained quantities are calculated by using both Monte Carlo schemes based on continuous slowing down approximation (CSDA) and on individual electron scattering events methods.